The following relates to the nuclear reactor arts, electrical power generation arts, and related arts.
Nuclear reactors employ a reactor core comprising a mass of fissile material, such as a material containing uranium oxide (UO2) that is enriched in the fissile 235U isotope. Primary coolant water, such as light water (H2O) or heavy water (D2O) or some mixture thereof, flows through the reactor core to extract heat for use in heating secondary coolant water to generate steam that is then used for some other useful purpose. For electrical power generation, the steam is used to drive a generator turbine. In thermal nuclear reactors, the water also serves as a neutron moderator that thermalizes neutrons, which enhances reactivity of the fissile material. Various reactivity control meachanisms, such as mechanically operated control rods, chemical treatment of the primary coolant with a soluble neutron poison, or so forth are employed to regulate the reactivity and resultant heat generation.
In a pressurized water reactor (PWR), the primary coolant water is maintained in a subcooled state in a sealed pressure vessel that also contains the reactor core. In the PWR, both pressure and temperature of the primary coolant water are controlled. One advantage of the PWR approach is that the primary coolant water, in which the radioactive reactor core is immersed, is contained within the pressure vessel.
For power generation, secondary coolant water is flowed in thermal communication with the primary coolant water. Heat (i.e., energy) is transferred from the reactor core to the secondary coolant water via the intermediary of the primary coolant water. This heat converts the secondary coolant water from liquid water to steam. The steam is typically flowed into a turbine or other power conversion apparatus that makes practical use of the steam power.
The apparatus providing thermal communication between the primary and secondary coolant water is referred to as the steam generator. In external steam generator designs, a large-diameter pressure vessel penetration carries primary coolant out of the pressure vessel into the external steam generator where it is flowed proximate to the secondary coolant water circuit. This approach has the disadvantage of taking primary coolant water outside of the pressure vessel, and of introducing a large-diameter vessel penetration for this purpose.
In internal steam generator designs, the steam generator is located (or “integrated”) inside the pressure vessel. In these “integral PWR” designs, secondary coolant water is flowed into the internal steam generator (and hence into the pressure vessel) and is heated to produce steam that flows out of the pressure vessel. Advantages include typically smaller pressure vessel penetrations, and avoiding flow of primary coolant water out of the pressure vessel.
However, the integral PWR approach has some disadvantages. Placing the steam generator inside the pressure vessel increases the size of the pressure vessel, which has disadvantages in terms of manufacturing cost, transport cost (e.g., a larger rail transport car and/or specialized rail system), site considerations (e.g., a deeper subterranean containment structure), and safety (e.g., more difficult to maintain seal integrity as the size of the pressure vessel increases). For various reasons including safety and primary coolant flow circuit design, it is conventional to place the nuclear reactor core at or near the bottom of the pressure vessel. Primary coolant water heated by the reactor core naturally tends to rise through the pressure vessel as it cools (the “hot leg” of the primary circuit), and the cooling primary coolant then flows downward to return to the reactor core (the “cold leg” of the primary circuit). In assisted circulation designs, this natural circulation is augmented by primary coolant pumps. The internal steam generator is placed above the reactor core, typically in the cold leg.
An illustrative example of such an integral PWR is shown in Thome et al., “Integral Helical Coil Pressurized Water Nuclear Reactor”, U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. This publication discloses a steam generator employing helical steam generator tubing; however, other coil geometries including straight (e.g., vertical) steam generator tubes are also known. This publication also discloses an integral PWR in which the control rod drive mechanism (CRDM) is also internal to the pressure vessel; however, external CRDM designs are also known. Some illustrative examples of internal CRDM designs include: Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2010/0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety; and Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, Intl Pub. WO 2010/144563 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety.
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